Volume 86, Issue 6

In the present study, CFD calculations are presented for the three types of water-based nanofluids Al 2 O 3 /water, CuO/water and TiO 2 /water with 0.1% volume fraction. These calculations are done with ANSYS-CFX and as geometry the SRBTL test loop as scaled down test loop for a VVER-1000 reactor core design is used. The goal of this study is to evaluate the CFD program against the SRBTL test loop core as a scaled core for applying water-based nanofluids as coolant. ANSYS-CFX simulation data are validated against the RELAP5/MOD3.2 simulation data for pure water. This comparison shows a good agreement. The simulation results for the nanofluids and water including Re number, temperature, viscosity, pressure drop and heat transfer coefficient through the SRBTL test loop core are compared. The results of the comparisons show that the SRBTL test loop core is suitable to extract experimental data of water-based nanofluids for using them as coolant in the VVER-1000 reactor.

This paper provides the empirical basis to support predictions of the Human Factor Engineering (HFE) influences in Human Reliability Analysis (HRA). A few methods were analyzed to identify HFE concepts in approaches of Performance Shaping Factors (PSFs): Technique for Human Error Rate Prediction (THERP), Human Cognitive Reliability (HCR) and Cognitive Reliability and Error Analysis Method (CREAM), Success Likelihood Index Method (SLIM) Plant Analysis Risk – Human Reliability Analysis (SPAR-H), A Technique for Human Error Rate Prediction (ATHEANA) and Man-Machine-Organization System Analysis (MMOSA). Also, in order to identify other necessary PSFs in HFE, an additional investigation process of human performance (HPIP) in event occurrences was used. Thus, the human error probability could be reduced and its evaluating can give out the information for error detection and recovery. The HFE analysis model developed using BHEP values (maximum and pessimistic) is based on the simplifying assumption that all specific circumstances of HFE characteristics are equal in importance and have the same value of influence on human performance. This model is incorporated into the PSA through the HRA methodology. Finally, a clarification of the relationships between task analysis and the HFE is performed, ie between potential human errors and design requirements.

The purpose of this study is to display the neutronic simulation of nanofluid application to reactor core. The variations of VVER-1000 nuclear reactor primary neutronic parameters are investigated by using different volume fraction of nanofluid as coolant. The effect of using nanofluid as coolant on reactor dynamical parameters which play an important role in the dynamical analysis of the reactor and safety core is calculated. In this paper coolant and fuel temperature reactivity coefficients in a VVER-1000 nuclear reactor with nanofluid as a coolant are calculated by using various volume fractions and different sizes of TiO 2 (Titania) nanoparticle. For do this, firstly the equivalent cell of the hexagonal fuel rod and the surrounding coolant nanofluid is simulated. Then the thermal hydraulic calculations are performed at different volume fractions and sizes of the nanoparticle. Then, using WIMS and CITATION codes, the reactor core is simulated and the effect of coolant and fuel temperature changes on the effective multiplication factor is calculated. For doing optimization, an artificial neural network is trained in MATLAB using the observed data. The different sizes and various volume fractions are inputs, fuel and coolant temperature reactivity coefficients are outputs. The optimal size and volume fraction is determined using the neural network by implementing the genetic algorithms. In the optimization, volume fraction of 7% and size 77 nm are optimal values.

Liquid organic scintillators are widely used in non-destructive analysis, which plays an important role in nuclear disarmament verification. This paper focused on studying the neutron-gamma discrimination technology in the fast neutron multiplicity measuring counter based on BC501A liquid scintillation detector. First, the charge comparison method, the zero-crossing time method and the rise time method were compared via the Geant4 and Matlab algorithm, and the result shows that charge comparison has the highest Figure of Merit. Then, a neutron-gamma discrimination system based on the six-probe fast neutron multiplicity counter was built and tested with a conclusion that the mean value of Figure of merit is 1.08, which verify the satisfactory neutron-gamma discriminating capability of the system. Finally, for the uranium samples, the mass are detected by fast neutron multiplicity counter, and the enrichment are measured by the characteristic gamma-ray signals using the system. The experimental results are in good agreement with the actual data.

The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.

In this study, we have proposed an analytical method for calculating the specific heat capacity of uranium nitride nuclear material. The specific heat capacity results have obtained by the use of the Debye-Einstein approximation. The thermal conductivity of nuclear material has been obtained by using the experimental data of thermal diffusivity and the calculation results of specific heat capacity. This method shows that our results are satisfactory for the wide range temperature variations. The proposed approach can be easily applied to determine the thermodynamic properties of the other nuclear materials.

This paper reviews an important integral test facility (ITF) named PKL (primary loop in German), which is designed based on a 4-loop pressurized water reactor (PWR) with the power 1 300 MWe, and especially concentrates on two aspects: (1) the tests at each developmental period of the facility until 2020, which is a typical microcosm of nuclear safety research; (2) the simulation of the PKL facility tests by using system thermal-hydraulic (STH) codes, especially RELAP5, TRACE and ATHLET. The results from the literature showed that all of these codes could reproduce the accident scenarios on the PKL facility to some extent, and simulate the complex phenomena both in the reactor pressurized vessel (RPV) and in the loops well, except some local phenomena (e. g., peak cladding temperature (PCT)). Furthermore, this paper presents some suggestions on PKL further tests. Especially, the sensitivity studies of initial conditions (ICs) and boundary conditions (BCs), test studies related to Extensive damage mitigation guidelines (EDMGs) and FLEX strategies, anticipated transients without scram (ATWS), detailed core section model, combination with other ITF or separate effects test (SET) facilities, and tests on advanced conception reactors are emphasized.

Oxygen and magnesium isotopes can be used in nuclear reactor materials as cooling, shielding, coating, electronics etc. They can also occur through nuclear reactions during the reactor operation. The exposure of high energy gamma can change the material and its properties, and hence its objective of selection may not remain satisfied. Thus, it is required to study the cross section of different reactions on nuclear reactor materials to understand their sustainability for the properties, for which they are chosen. In the scope of this study, theoretically, different level density model calculations and γ-ray strength functions have been performed for (γ, p) reaction for 16,18 O and 24,26 Mg nuclei using TALYS 1.9 and EMPI˙RE 3.2.2 codes. Also, semi empirical (γ, p) formula by Tel et al., have been calculated and compared with all results. The effect of different level density models defined in these codes on gamma strength has been studied. Finally, the consistency of these obtained data with EXFOR data have been investigated.

In this study, 90 Sr was used as the test radionuclide to characterize the sorption kinetics and effects of initial 90 Sr activity and remaining 90 Sr in solid concentration were simulated for a near-surface repository. The study focused on the sorption characteristics of radionuclides in unsaturated groundwater environment (or vadose zone) is the important information for investigating the near-surface disposal of intermediate and low-level radioactive waste (ILLW). Moreover, the 90 Sr sorption experiments reached equilibrium within 56 h, which fit to the first order sorption kinetic model, and the remaining 90 Sr in mudrock samples showed obvious sorption equilibrium hysteresis, which fit to the second order sorption kinetic model. Before reaching the maximum sorption capacity, the sorption rate constant increases with 90 Sr increasing; the distribution coefficient (K d ) of 56 h decreases with the remaining 90 Sr decreasing. In addition, it showed that the slow sorption process dominated before the sorption reaches equilibrium. In fact, a reliable safety assessment methodology for on-going near-surface repository required a lot of the radionuclides parameters with local environment including the radionuclides sorption/desorption rate constant and maximum sorption capacity.